1. Field
The present invention relates generally to conduit length adjustment apparatuses for steam generators and more particularly to a conduit length adjustment apparatus between a probe pusher and a steam generator tube sheet in a steam generator.
2. Related Art
Pressurized water nuclear reactors employ steam generators to isolate and place a radioactive coolant, flowing in a primary circulation loop, in heat exchange relationship with a secondary fluid flowing in a secondary circulation loop. Steam is generated from the secondary fluid. The steam generally is employed to drive a turbine to perform work, e.g., an electric generator. In the primary loop the reactor coolant is heated by the nuclear reactions occurring in the reactor core and circulated through a hot piping leg to a hemispherical bowl shaped portion of the primary side of the steam generator generally known as the channel head. The channel head is separated, by a partition across its diameter, into inlet and outlet plenums, which are covered by a tube sheet through which the terminating ends of U-shaped heat exchanger tubes are fastened. Each of the U-shaped heat exchanger tubes originate in a bore in the tube sheet passing from the inlet plenum of the channel head and terminate in a bore in the tube sheet that communicates with the outlet plenum of the channel head. A cylindrically shaped secondary side of the steam generator is disposed around and over the tube sheet and the U-shaped heat transfer tubes. Hot, radioactive water from the reactor core circulates through the primary side of the steam generator, while non-radioactive water is introduced into the secondary side. The tube sheet and heat exchanger tubes hydraulically isolate but thermally connect the primary side to the secondary side. Hot radioactive water from the primary side flows through the interior of these heat exchanger tubes while the exterior of these tubes come into contact with the non-radioactive water in the secondary side in order to generate nonradioactive steam.
In the secondary side of the steam generator exterior portions of the U-shaped heat exchanger tubes are supported by and extend through bores present in a plurality of horizontally supported plates that are vertically spaced along the elongated length of the tubes. Small annual spaces are present between the heat exchanger tubes and the bores in the support plates, and the tube sheet, which are known in the art as “crevice regions.” Such crevice regions provide only a very limited flow path for the feed water that circulates throughout the secondary side of the steam generator, which causes “dry boiling” to occur wherein the feed water boils so rapidly that these regions can actually dry out during operation of the steam generator. This chronic drying out causes impurities in the water to precipitate and collect in these crevice regions. These precipitates ultimately create sludge and other debris that promotes the occurrence of corrosion in the crevice regions which, if not repaired, can ultimately cause the tube to crack and to allow radioactive water from the primary side to contaminate the non-radioactive water in the secondary side of the steam generator.
Eddy current probe systems are employed to monitor the extent of degradation in the walls of the heat exchanger tubes that result from corrosion. One such system is described in U.S. Pat. No. 5,174,165 issued Dec. 29, 1992 to the assignee hereof. One of the services performed at a nuclear power plant is eddy current inspection of the steam generator tubing using such a system. The inspection involves insertion and removal of various configurations of eddy current probes in the high radiation and contaminated area of a nuclear steam generator. Minimizing personal time and equipment near the manway opening through which access to the interior of the steam generator is obtained (generally referred to as the steam generator platform) is highly desirable due to the elevated radiation level in that area. Typically the probes are attached to a long flexible piece of tubing (poly) and driven with a probe pusher through a flexible conduit to an area of interest or the entire length of the steam generated tube. One end of the flexible conduit is generally fixed to the probe pusher while the opposite end is attached to and positioned under the steam generator tube with a robotic manipulator.
A problem during eddy current inspection is that the amount of conduit in the steam generator needs to be increased or decreased as the robotic manipulator moves to various tube locations. This task is typically accomplished by manually adding or removing sections of the flexible conduit on the steam generator platform, which is a source of radiation exposure time for the field service operators. One method that does not require conduit length change is described in U.S. Pat. No. 6,606,920 issued Aug. 19, 2003 to the assignee hereof U.S. Pat. No. 6,606,920 describes a system in which the probe pusher is mounted to a drive system which enables both the probe pusher and the conduit to reposition during eddy current inspection. While the arrangement described in U.S. Pat. No. 6,606,920 is effective, the amount of working space required to translate the probe pusher is not available at many power plants.
It is an object of this invention to overcome these difficulties.